Free acldlty In uranlum(V1) and plutonium( IV) solutions has been determlned with sodium sulfate as complexant and sodlum carbonate as tltrant. The described method ls simple, accurate, and appllcable to all ranges of nHric acld and heavy metal concentratlons relevant to Purex process. The method Is adaptable for remote Operation. The overall recovery of nitrlc acid Is 99.5 % with 0.7 % relative standard deviation.Uranlum content has also been determined In the same allquot wlth a recovery of 98.5% and 1.2% relative standard deviation after the determlnatlon of acldlty.Analytical procedures for the determination of free acidity in uranium(V1) and plutonium(1V) solutions are required for nuclear fuel reprocessing. Knowledge of distribution ratios of actinides and nitric acid in extraction systems involving tri-n-butyl phosphate (TBP) and an alkane diluent is of great importance for modeling of the extraction or reextraction step of the Purex process. Accurate determination of free nitric acid is necessary for solvent extraction, ion exchange, and precipitation operations that separate and purify the actinide products. Free acid content must be established prior to neutralization of nitric acid in waste solutions generated from these operations. The knowledge of free nitric acid concentration is also required in the studies of third phase formation and complex formation. However, the acidity produced by the hydrolysis of uranium and plutonium due to their large ionic potentials contributes to the initial acidity of the solution and poses problem in the accurate determination of free acidity. To remove hydrolytic interferences of uranium and plutonium and to prevent polymerization and disproportionation reactions of plutonium, these species are chemically removed by adding complexing agents. A survey of literature reveals that the most commonly used complexing agents are oxalate (1, 2), citrate (3), fluoride (4,5), EDTA (6), or a mixture of fluoride and oxalate (7,8). The use of neutral potassium oxalate as complexing agent for uranium has been reported to give biased results with the magnitude of bias depending on the concentration and species of hydrolyzable ions present and with the amount of oxalate salt used. The potassium oxalate method is not reliable a t high uranium to nitric acid ratios. In the case of plutonium, complexation with oxalate and subsequent titration to a preselected end point (pH 5.55) have been reported to result in an acceptable bias over a restricted acid range centered around 1 mequiv of acid. Therefore, the neutral potassium oxalate method serves only as a relative index for plant acidity control and cannot be accepted as an accurate method. Other complexing agents have also been used with varying degrees of success with neutralization titrations. Other methods based on precipitation and separation of heavy metals followed by alkalimetric determination (9) and methods invovling the partial neutralization of acids by standard addition of base and the calculation of end point by Gran plot ...

The organic compounds formed during the dissolution of carbide fuels in nitric acid adversely affect the extraction of plutonium by tributyl phosphate. This work was undertaken as a preliminary study of the effect of refluxing the dissolver solutions obtained by dissolution of uranium carbide in 13 Μ nitric acid on the subsequent solvent extractions steps. The extraction of plutonium from carbide dissolver solutions by 30% TBP in ndodecane and the subsequent stripping and partitioning of Pu from the organic phase loaded with uranium and plutonium were studied. The results obtained were compared with those obtained with control solutions containing uranyl nitrate and plutonium(IV) nitrate with concentrations of uranium, plutonium and nitric acid similar to those of the dissolver solutions. The distribution ratios obtained with the dissolver solution and the control solution were comparable, indicating that the dissolver solution obtained after refluxing for 20h may give a feed solution suitable for PUREX process. Similarly, extraction and partitioning experiments were carried out with (U, Pu)C dissolver solutions and the results were similar to the results obtained with uranium carbide dissolver solutions.As an alternative chemical route for the destruction of carbonaceous materials in the UC dissolver solution, the effectiveness of sodium dichromate as an oxidizing agent was also investigated in terms of the percentage carbon remaining after treatment of the dissolver solution with sodium dichromate. The treatment resulted in destruction of nearly 97% of the initial carbon present in the carbide fuel.

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Yttrium oxide (Y2O3) thin films were deposited by microwave electron cyclotron resonance (ECR) plasma assisted metal organic chemical vapour deposition (MOCVD) process using indigenously developed metal organic precursors Yttrium 2,7,7‐trimethyl‐3,5‐octanedionates, commonly known as Y(tod)3 which were synthesized by an ultrasound method. A series of thin films were deposited by varying the oxygen flow rate from 1–9 sccm, keeping all other parameters constant. The deposited coatings were characterized by X‐ray photoelectron spectroscopy, glancing angle X‐ray diffraction and infrared spectroscopy. Thickness and roughness for the films were measured by stylus profilometry. Optical properties of the coatings were studied by the spectroscopic ellipsometry. Hardness and elastic modulus of the films were measured by nanoindentation technique. Being that microwave ECR CVD process is operating‐pressure‐sensitive, optimum oxygen activity is very essential for a fixed flow rate of precursor, in order to get a single phase cubic yttrium oxide in the films. To the best of our knowledge, this is the first effort that describes the use of Y(tod)3 precursor for deposition of Y2O3 films using plasma assisted CVD process.

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